Problems:

$6-$ Based on the $0\mathrm{.}025-eV$ cross section and decay constants in Fig $2-16($slides $14$&$15)$ and a thermal$-$neutron flux of $5\frac{{10}^{13}}{c}{m}^2-.$ Use the following cross section data.

$\backslash$table$[[$Nuclide$,$capture,fission$],[$U$-235,104b,586b$